1. Field
Example embodiments generally relate to apparatuses and methods for structurally replacing cracked welds. Example embodiments also relate to nuclear power plants and to apparatuses and methods for structurally replacing cracked welds of the nuclear power plants.
2. Description of Related Art
In many applications, such as nuclear reactors, steam driven turbines, or water deaerators, high-temperature water may adversely affect the associated structures by contributing to stress corrosion cracks, corrosion, erosion, and so forth. For example, high temperature waters may contribute to stress corrosion cracking (“SCC”) in materials, such as carbon steels, alloy steels, stainless steels, nickel-based alloys, cobalt-based alloys, and zirconium-based alloys. SCC may preferentially occur with certain combinations of alloys, environment, and stress.
As would be understood by a person having ordinary skill in the art (“PHOSITA”), SCC may include cracks propagated by static or dynamic tensile stresses acting in combination with corrosion at crack tips. These stresses may result or originate from differences in thermal expansion or contraction between components, relatively high or varying operating pressures, or various processes performed during the manufacture and assembly of the components or system. For example, residual stresses often result from cold working, grinding, machining, and other thermo-mechanical metal treatments. Water chemistry, welding, heat treatment, and radiation may also increase the susceptibility of metal or alloy component to SCC. SCC may be transgranular or intergranular in nature.
SCC may occur at greater rates under various conditions, such as the presence of oxygen, high radiation flux, and so forth. In nuclear reactors such as a pressurized water reactor (“PWR”) or a boiling water reactor (“BWR”), high radiation flux may cause radiolytic decomposition of the reactor coolant (water); this decomposition may produce oxygen, hydrogen peroxide, short-lived radicals, and various oxidizing species. These products of radiolytic decomposition may promote SCC in the various system components, such as pipes, pumps, valves, turbines, and so forth. Operating temperature and pressure for a BWR may be about 300° C. and about 10 MPa, and those for a PWR may be about 325° C. and about 15 MPa. Thus, the operating environment for BWRs and PWRs may increase the risk of having SCC issues in nuclear reactor components.
The microstructure of metals and alloys may include grains separated by grain boundaries. Intergranular stress corrosion cracking (“IGSCC”) may be a more localized SCC attack along or adjacent to grain boundaries, with the bulk of the grains themselves remaining largely unaffected. IGSCC may be associated with chemical segregation effects (e.g., impurity enrichment at grain boundaries) or with specific phases precipitated at grain boundaries.
Irradiation assisted stress corrosion cracking (“IASCC”) may refer to acceleration of SCC by irradiation (e.g., irradiation-induced changes that may involve microstructure changes, microchemical changes, and compositional changes by transmutation). IASCC may result from the effects of beta radiation, gamma radiation, neutron radiation, or other particle radiation (e.g., ions). However, for BWRs and PWRs, IASCC may be primarily due to neutron radiation.
Due to the serious nature of IASCC, the Nuclear Regulatory Commission (“NRC”) commissioned a series of studies over about a ten-year period. Some of the reports coming out of these studies included NUREG/CR 5608, “Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steels Irradiated in the Halden Reactor”; NUREG/CR-6892, “Fracture Toughness and Crack Growth Rates of Irradiated Austenitic Stainless Steels”; NUREG/CR-6687, “Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steel Alloys”; NUREG/CR-6915, “Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels and Alloy 690 from Halden Phase-II Irradiations”; NUREG/CR-6960, “Crack Growth Rates and Fracture Toughness of Irradiated Austenitic Stainless Steels in BWR Environments”; and NUREG/CR-7018, “Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments”.
FIG. 1 is a sectional view, with parts cut away, of reactor pressure vessel (“RPV”) 100 in a related art BWR.
During operation of the BWR, coolant water circulating inside RPV 100 may be heated by nuclear fission produced in core 102. Feedwater may be admitted into RPV 100 via feedwater inlet 104 and feedwater sparger 106 (a ring-shaped pipe that may include apertures for circumferentially distributing the feedwater inside RPV 100). The feedwater from feedwater sparger 106 may flow down through downcomer annulus 108 (an annular region between RPV 100 and core shroud 110).
Core shroud 110 may be a stainless steel cylinder that surrounds core 102. Core 102 may include a multiplicity of fuel bundle assemblies 112 (two 2×2 arrays, for example, are shown in FIG. 1). Each array of fuel bundle assemblies 112 may be supported at or near its top by top guide 114 and/or at or near its bottom by core plate 116. Top guide 114 may provide lateral support for the top of fuel bundle assemblies 112 and/or may maintain correct fuel-channel spacing to permit control rod insertion.
The feedwater/coolant water may flow downward through downcomer annulus 108 and/or into core lower plenum 118. The coolant water in core lower plenum 118 may in turn flow up through core 102. The coolant water may enter fuel assemblies 112, wherein a boiling boundary layer may be established. A mixture of water and steam may exit core 102 and/or may enter core upper plenum 120 under shroud head 122. Core upper plenum 120 may provide standoff between the steam-water mixture exiting core 102 and entering standpipes 124. Standpipes 124 may be disposed atop shroud head 122 and/or in fluid communication with core upper plenum 120.
The steam-water mixture may flow through standpipes 124 and/or may enter steam separators 126 (which may be, for example, of the axial-flow, centrifugal type). Steam separators 126 may substantially separate the steam-water mixture into liquid water and steam. The separated liquid water may mix with feedwater in mixing plenum 128. This mixture then may return to core 102 via downcomer annulus 108. The separated steam may pass through steam dryers 130 and/or may enter steam dome 132. The dried steam may be withdrawn from RPV 100 via steam outlet 134 for use in turbines and other equipment (not shown).
The BWR also may include a coolant recirculation system that provides the forced convection flow through core 102 necessary to attain the required power density. A portion of the water may be sucked from the lower end of downcomer annulus 108 via recirculation water outlet 136 and/or may be forced by a centrifugal recirculation pump (not shown) into a plurality of jet pump assemblies 138 (only one of which is shown) via recirculation water inlets 140. Jet pump assemblies 138 may be circumferentially distributed around core shroud 110 and/or may provide the required reactor core flow.
As shown in FIG. 1, a related art jet pump assembly 138 may include a pair of inlet mixers 142. A related art BWR may include 16 to 24 inlet mixers 142. Each inlet mixer 142 may have an elbow 144 welded to it that receives water from a recirculation pump (not shown) via inlet riser 146. An example inlet mixer 142 may include a set of five nozzles circumferentially distributed at equal angles about the axis of inlet mixer 142. Each nozzle may be tapered radially inwardly at its outlet. Jet pump assembly 138 may be energized by these convergent nozzles. Five secondary inlet openings may be radially outside of the nozzle exits. Therefore, as jets of water exit the nozzles, water from downcomer annulus 108 may be drawn into inlet mixer 142 via the secondary inlet openings, where it may be mixed with coolant water from the recirculation pump. The coolant water then may flow into diffuser 148.
FIG. 2 is a schematic diagram showing a developed azimuthal view of the interior of a related BWR core shroud that comprises a plurality of shell sections, having vertical seam welds, that are welded together, one upon the next, by horizontal seam welds.
As shown in FIG. 2, core shroud 200 may comprise first shell sections 202a and 202b, second shell sections 204a and 204b, third shell sections 206a and 206b, fourth shell sections 208a and 208b, and fifth shell sections 210a, 210b, and 210c. Core shroud 200 may be supported by shroud supports 212a, 212b, and 212c, as well as shroud support plate 214.
Shroud supports 212a, 212b, and 212c may be joined together using vertical seam welds V12, V13, and V14, and also may be joined using horizontal seam weld H8 to shroud support plate 214.
Fifth shell sections 210a, 210b, and 210c may be joined together using vertical seam welds V9, V10, and V11 to form a lower shell section of core shroud 200, and also may be joined using horizontal seam weld H7 to shroud supports 212a, 212b, and 212c. 
Fourth shell sections 208a and 208b may be joined together using vertical seam welds V7 and V8 to form a bottom mid-core shell section of core shroud 200, and also may be joined using horizontal seam welds H6A and H6B to fifth shell sections 210a, 210b, and 210c. Horizontal seam weld H6A may represent the joining of fourth shell sections 208a and 208b to core plate support ring 216; horizontal seam weld H6B may represent the joining of core plate support ring 216 to fifth shell sections 210a, 210b, and 210c. 
Third shell sections 206a and 206b may be joined together using vertical seam welds V5 and V6 to form a middle mid-core shell section of core shroud 200, and also may be joined using horizontal seam weld H5 to fourth shell sections 208a and 208b. 
Second shell sections 204a and 204b may be joined together using vertical seam welds V3 and V4 to form a top mid-core shell section of core shroud 200, and also may be joined using horizontal seam weld H4 to third shell sections 206a and 206b. 
First shell sections 202a and 202b may be joined together using vertical seam welds V1 and V2 to form an upper shell section of core shroud 200, and also may be joined using horizontal seam welds H2 and H3 to second shell sections 204a and 204b. Horizontal seam weld H2 may represent the joining of first shell sections 202a and 202b to top guide support ring 218; horizontal seam weld H3 may represent the joining of top guide support ring 218 to second shell sections 204a and 204b. 
Horizontal seam weld H1 may represent the joining of shroud flange 220 to first shell sections 202a and 202b. 
As known to a PHOSITA, the relative offsets in vertical seam welds V1-V14 attempt to ensure that a crack in a single vertical seam weld cannot propagate over a significant distance (e.g., all the way from horizontal seam weld H1 to horizontal seam weld H8). However, horizontal seam weld H1-H8 do not have such an offset arrangement.
Although SCC, IGSCC, and IASCC have been studied, no “cure” has been found. As a result, cracks continue to initiate and propagate in components of nuclear reactors. Core shrouds may be particularly susceptible due to their extremely high neutron fluence as the nuclear reactor ages. For example, in core shroud 200, the active fuel in an associated core 102 may extend vertically from between horizontal seam welds H5 and H6A to about horizontal seam weld H2 or H3. Thus, horizontal seam welds H2, H3, H4, and H5, and vertical seam welds V3, V4, V5, V6, V7, and V8, all may be described as being subject to extremely high neutron fluence. In the event of SCC, IGSCC, or IASCC of the seam welds, core shroud 200 could be replaced. However, a more economically feasible approach might be to conduct weld repair or to structurally replace the horizontal seam welds, vertical seams welds, or both.
Such a weld repair may be done with the seam welds submerged, but this approach may be difficult from a technical point of view. Such a weld repair also may be done with the seam welds not submerged, but this approach may present other issues, such as significant radiation exposure and extension of the critical path during an outage.
As known to a PHOSITA, tie-rods have been proposed to structurally replace the horizontal seam welds as a group. Although tie-rods may provide significant support for the horizontal seam welds as a group, such tie-rods may not be as effective in structurally replace individual horizontal seam welds.
As also known to a PHOSITA, various devices have been proposed to structurally replace the vertical seam welds. Most of these devices involved full penetration of the structure that includes the vertical seam welds. Although such devices may be employed, full penetration of the structure that includes the vertical seam welds may introduce other problems, such as creating potential leakage paths, complicating installation procedures and reactor safety calculations, and establishing new periodic inspection requirements.
Thus, a need exists for apparatuses and methods that may provide the ability to structurally replace individual welds in nuclear reactor components subject to SCC, IGSCC, or IASCC. In the case of core shroud 200, this may include structurally replacing individual horizontal seam welds, individual vertical seams welds, or both. In particular, a need exists for apparatuses and methods that may provide the ability to structurally replace individual welds in nuclear reactor components subject to SCC, IGSCC, or IASCC without fully penetrating a structure that includes the individual welds.
Related art systems, methods, and/or filters for apparatuses and methods for structurally replacing cracked welds are discussed, for example, in U.S. Pat. No. 5,392,322 to Whitling et al. (“the '322 patent”); U.S. Pat. No. 5,521,951 to Charnley et al. (“the '951 patent”); U.S. Pat. No. 5,530,219 to Offer et al. (“the '219 patent”); U.S. Pat. No. 5,538,381 to Erbes (“the '381 patent”); U.S. Pat. No. 5,621,778 to Erbes (“the '778 patent”); U.S. Pat. No. 5,675,619 to Erbes et al. (“the '619 patent”); U.S. Pat. No. 5,712,887 to Thompson et al. (“the '887 patent”); U.S. Pat. No. 5,729,581 to Loock et al. (“Loock”); U.S. Pat. No. 5,737,379 to Erbes (“the '379 patent”); U.S. Pat. No. 5,742,653 to Erbes et al. (“the '653 patent”); U.S. Pat. No. 5,802,129 to Deaver et al. (“the '129 patent”); U.S. Pat. No. 5,803,686 to Erbes et al. (“the '686 patent”); U.S. Pat. No. 5,803,688 to Gleason et al. (“the '688 patent”); U.S. Pat. No. 6,067,338 to Erbes (“the '338 patent”); U.S. Pat. No. 6,138,353 to Weems et al. (“Weems I”); U.S. Pat. No. 6,343,107 B1 to Erbes et al. (“the '107 patent”); U.S. Pat. No. 6,345,927 B1 to Pao et al. (“the '927 patent”); U.S. Pat. No. 6,371,685 B1 to Weems et al. (“Weems II”); U.S. Pat. No. 6,464,424 B1 to Weems et al. (“Weems III”); and U.S. Pat. No. 7,649,970 B2 to Erbes (“the '970 patent”); and in U.S. Patent Publication No. 2003/0234541 A1 to Thompson et al. (“the '541 publication”); U.S. Patent Publication No. 2011/0101177 A1 to Suganuma et al. (“Suganuma I”); and U.S. Patent Publication No. 2012/0087456 A1 to Suganuma et al. (“Suganuma II”).
The disclosures of the '107 patent, the '129 patent, the '219 patent, the '322 patent, the '338 patent, the '379 patent, the '381 patent, the '619 patent, the '653 patent, the '686 patent, the '688 patent, the '778 patent, the '887 patent, the '927 patent, the '951 patent, and the '970 patent are incorporated in this application by reference in their entirety. Similarly, the disclosures of the '541 publication are incorporated in this application by reference in its entirety.